1. Field of the Invention
This invention relates to a field of radiochemical technology, more specifically to radioactive waste reprocessing, and can be used for extraction recovery of metals.
2. Description of the Related Art
Ecological safety requires that long-lived radionuclides be transferred into forms preventing their distribution into the environment during storage. To reduce the volume of radioactive waste and thus to make its storage cost effective, the major radionuclides, e.g., cesium (Cs), strontium (Sr), and actinide elements (An) can be separated from stable elements (sodium, aluminum, etc.) Depending on specific requirements, it may be necessary to recover all the major radionuclides or certain fractions of them. The most common method for highly radioactive liquid waste (HAW) treatment is liquid-liquid extraction. The extraction process consists of the extraction operation whereby the radionuclides (An, Cs, Sr, etc.) are transferred into the organic phase and the stripping operation whereby the radionuclides pass back into an aqueous phase. The resultant strip product (aqueous radionuclide-containing solution) is thereafter subjected to immobilization, usually by vitrification. To reduce the volume of glass produced by vitrification, the salt content in the resultant strip product should be minimized.
The method for stripping uranium from the organic phase consisting of tributylphosphate in kerosene with the use of soda solution (e.g., Na2CO3) is known. See J. C. Mailer and O. K. Tallent, “A Review of Recent ORNL Studies in Solvent Cleanup and Diluent Degradation,” paper presented at US/UK Exchange Meeting on Solvent Extraction Technology, Oak Ridge, Tenn., Oct. 25-28, 1982; and Goldacker, H. et al., Kerntechnik, v.18, p. 426 (1976). The method for stripping americium from the organic phase with the use of soda solution (e.g., Na2CO3) and complexant is known, see P. S. Dhami et al., Sep. Sci. Technol., v36(2), p. 325-335, (2001). The resulting strip products contains a considerable amount of sodium.
The methods for radionuclide stripping from organic solutions based on chlorinated cobalt dicarbollide (ChCoDiC) with synergistic additives are known. See J. Rais, S. Tachimori, P. Selucky and L. Kadecova, “Synergistic Extraction in Systems with Dicarbollide and Bidentate Phosphonates,” Sep. Sci. Technol. 29 (2), 261-274, 1994; J. Rais and L. Kadlecova, “Method of the Extraction Isolation of Tervalent Lanthanides and Actinides from Aqueous Solutions,” Czechoslovakian Patent 216,101 (November 1984); J. Rais, M. Kyrs, P. Selucky, “Method of the Extraction Isolation of Strontium from Aqueous Solutions,” Czechoslovakian Patent 224,890 (October 1985); J. Rais, M. Kyrs, S. Hermanek, “Method of Sr Isolation from Aqueous Solution,” Czechoslovakian Patent 153,993 (June 1974). Both HNO3 and mixtures of HNO3 with organic reagents like hydrazine nitrate, amine (methylamine, dimethylamine) nitrates, dimethylformamide, trimethylphosphate, etc. are also used as stripping solutions. By using all the proposed methods it is possible to obtain a strip product containing minimal salts of stable metals. However, the methods listed above give strip products containing considerable quantities of nitric acid or mixture of nitric acid and organic compounds that require destruction by additional chemical reagents or by oxidation at elevated temperature.
The claimed method is most similar to that for the stripping of cesium and other radionuclides from organic solution by a solution of guanidine carbonate and DTPA. See V. N. Romanovskiy, I. V. Smirnov, V. A. Babain, T. A. Todd, J. D. Law, R. S. Herbst, and K. N. Brewer, “The Universal Solvent Extraction (UNEX) Process I: Development of the UNEX Process Solvent for the Separation of Cesium, Strontium, and the Actinides from Acidic Radioactive Waste,” Solvent Extraction and Ion Exchange, 19 (1), pp. 1-21 (2001). In accordance with the the method of stripping used in the UNEX process, the organic solution of 0.05-0.08 M ChCoDiC+0.007-0.02 M PEG-400+0.01-0.02 M diphenyl-N,N-dibutylcarbomoyl-phosphine oxide (CMPO) containing Cs, Sr, Am, and Pu, is brought into contact with the solution of 0.5-1.0 M guanidine carbonate+0.02-0.04 M diethylenetriaminepenta-acetic acid (DTPA). In this case, the stripping distribution coefficients are 0.2-0.3 for Cs and below 0.01 for other metals.
The drawback of the UNEX method is that the resulting strip product contains considerable amounts of organic compounds: 90-200 g/L guanidine carbonate and 10-20 g/L diethylenetriaminepentaacetic acid (DTPA). For subsequent vitrification, the strip product can be preliminarily dried and the guanidine destroyed. The process of guanidine destruction requires large amounts of oxidizer and should be conducted with heating. Oxidizing and high-temperature unit operations, with highly radioactive solutions of the strip product, are complicated technical tasks, and raise safety issues.
While the UNEX process of stripping (and the UNEX process) is a viable extraction/stripping method for nuclear wastes that are acidic (due to presence of nitric acid), extraction systems based on alkylated cobalt dicarbollides in combination with PEG's are promising extractants for Cs and Sr recovery from highly basic nuclear waste solutions. See J. Rais, P. Selucky, N. V. Sistkova, and J. Alexova, “Extraction of 137Cs and 90Sr from Alkaline Solutions with High NaNO3 Content with Tetrahexyldicarbollide,” Sep. Sci. Technol. 34 (14), 2865-2886, 1999; R. M. Chamberlin and K. D. Abney, “Strontium and Cesium Extraction into Hydrocarbons Using Alkyl Cobalt Dicarbollide and Polyethylene Glycols”, J. Radioanalytical & Nuclear Chem., Vol. 240, No. 2, 547-553, May 1999. Since the alkylated derivatives of cobalt dicarbollide used in the extraction of Cs and Sr from highly basic wastes are not stable in acidic medium, the necessity for stripping with basic solutions is mandatory. The claimed method of stripping is compatible with and applicable to use with these basic-side extraction systems.